This invention pertains generally to protection systems for nuclear reactors, and more particularly, to protection systems employing setpoints variably dependent upon the prior core history, to control reactor operation.
Generally, nuclear reactors contain a reactive region commonly referred to as the core in which sustained fission reactions occur to generate heat. The core includes a plurality of elongated fuel rods comprising fissile material, positioned in assemblies and arranged in a prescribed geometry governed by the physics of the nuclear reaction. Neutrons bombarding the fissile material promote the fissionable reaction which in turn releases additional neutrons to maintain a sustained process. The heat generated in the core is carried away by a cooling medium, which circulates among the fuel assemblies and is conveyed to heat exchangers which in turn produce steam which forms the motive force to drive turbine generators for the production of electricity.
Commonly, in pressurized water reactors a neutron absorbing element is included within the cooling medium (which also functions as a moderator) in controlled variable concentrations to modify the reactivity when required, and thus the heat generated within the core. In addition, control rods are dispersed among the fuel assemblies, longitudinally movable axially within the core, to control the core's reactivity, and thus its power output.
While the radial power distribution of the core is fairly uniform under normal operation, due to the prescribed arrangement of fuel assemblies and control rods which are symmetrically situated radially throughout the core, the axial power distribution can vary greatly during reactor operation. Preferably, to obtain maximum efficiency in fuel burnup and retain a maximum power output capability within the core, the axial power distribution is maintained substantially uniform under most operating conditions.
The neutron flux within the core is monitored as a representation of core power, by four axially spaced detectors equidistantly positioned around the periphery of the core, exterior of the reactor. Each detector monitors the flux in the upper and lower half of a corresponding core quadrant and provides corresponding outputs which are employed by the protection and control systems of the reactor. Flux control limits are established to assure that potential axial and radial flux peaks are maintained within acceptable limits.
One of the protection systems offered for pressurized water reactors trips the reactor and ceases the core fission reaction when the flux detectors identify a negative rate of change of flux within the core greater than a preestablished value. Such a negative rate of change of flux can, for example, be indicative of a dropped control rod, which will alter the radial flux symmetry within the core and reduce the overall core power output, and thus the heat generated by the core. Without such protection the programmed average temperature control system employed in a number of nuclear electrical generating facilities would attempt to increase the core power output automatically upon such a reduction in power, to maintain load requirements, without consideration of the radial power symmetry of the core. An example of such a control system is described in U.S. Pat. No. 3,423,285 to C. F. Currey et al. Such an increase in power level without consideration of the core power symmetry and the heat operating power level within the core could raise local core power conditions above acceptable limits. Furthermore, the reduced power output of the core, without a corresponding reduction in load, will reduce the temperature of the reactor. In nuclear systems having a negative temperature coefficient, such as in pressurized water reactors, the lower core temperature will result in an increase in reactivity which can also raise local conditions above desirable limits. However, it is not always necessary to trip a reactor under such conditions if local power peaks can be maintained below design limits.
Accordingly, a new protection and control system is desired that will identify abnormal operating conditions and modify reactor control compatibly with continued, safe operation of the plant.